ML20141K620
Text
{{#Wiki_filter:m. V;. 1 (; u ' RECEIVED l + gverhe,r
- m. m.y 20, 1968.
J ARGON N L: NATIONAL LABORATORY j Sd vP) T Mr. R.-D. Young,-Chief FMq h -l Containment:& Component Tech. Branch g4 ' Division of Reactor Licensing g3 ,DU U.~ Si Atomic Energy Commission ' Washington, D. C. 20545
Dear Mr. Young:
Subject:
'Intergranular Cracking-- Oyster Creek Reactor Vessel i
n-Thic memorandum is a materials evaluation en the corrosion aspect of the .l cracked vessel components which have been subsequent 24, 1968.11}y rapaired. It super-cedes the previous memorandum dated July I. Conclusions & Recommendations 1 A. -When Type 304 stainless steel is sensitized, it appears the crack- 'ing vill progress more rapidly than in solution annealed condition und will be predominantly intergranular. Whether the. resultant cracking is in fact stress-assisted intergranular corrosior,or indeed stress corrnaion cracking -is unresolved. The reason is both cracking mechanisms are not clearly j understood. i 3. In the case of. internal vessel components (e.g. shroud support ring) -any.intergranular cracking that developes under normal operating conditions is expected to be a ductile fracture.- Brittle fracture and re- ] lease of small lumps of material to the coolant are not anticipated. C.- In the case ofl exterval vessel components (e.g. : recirculating nozzle safe ends), any. surface.intergranular cracking, due te, unsuspected contaminants can be detected during periodic inspection or_ scheduled ^ shut-- downs. -0 D. If:no inspection of external vessel components is: contemplated: ' then outsidc vessel monitoring program of r?,ressed specicens of Type 3081. L
- stainless steel veld overlay is ;dvisable.
Similar specimens ofLType 204 stainless steel in solution-annealed and sensttized conditions can be used as controls., i II. - Vessel Component Cracking and Repair F*stus. 'Intergranular_ corrosion'and. weld defects were found Ln stub tubes of [ the primary pressure vessel at Oyster Creek-BWR Power Plant after the field ~ [AN8?N 3) ,e V 4f ).5f 7.?d424t2 n04 60439
- ibidphhne $1 -
i X 910-258 3282 VdUX LB. Argonne. Illinois 9700 South s nue, ~ nn .u.- 32 2'9$oso7 DEKOK91-282 PDR p. ~ ... ~ -, --.
( Mr. R. D. Young - November 20, 1068 hydrostatic test in October, 1967.(2) Further examination revealed similar intergranular cracking on the ID and OD surfaces of the safe ends of the cora spray nozzle, the control rod drive return nozzle, and on the isolation con-denser norries. In addition, the same type of failures were found on the OD surface of the vessel head nozzle flanges and the recirculation nozzle safe ~ ends. However, no indications of intergranular cracking were found on the ID of the stub tubes, the ID)of the main recirculation nozzle safe ends, nor the ID of the head nozzles.C3 All the above centioned f ailed components (which had been furnace sen-sitized) were either surface overlayed with 308L weld material or replaced. Subsequant dye penetrant examination revealed no indication of failure in both before and after hydrotest. N III. Shop and Site Environpent Investigations l Combustion Engineering has produced intergranular cracking in stressed furnace sensitized 304 stainless steel af ter contaminating the material with carbon steel prior to heat treating, followed by sequential exposure to simu-lated shop and field hydro test solutions and field cleaning solutions. In-tergranular attack has also been produced in aqueous chloride solutions, and in gluconic acid solutions which represent one of the oxidation products of the dextrose used in the EDTA chalating step. - However,.the cracking observed in the sequential exposures at Combustion Engineering has not be,en observed in single exposures to the same solutions at General Electric. Confirmation tests of :his particular sequence are in pragress a t General Electric. In addition, tests are being performed in which iron uxide is present as a sus-4 pension in tPe solution.5 1 However, work performed in this continuing in-vestigation has not clearly identified the cause of the attacklin the Oyster Creek vessel. S'hould be pointed out that G.E.'s statement on page 3-15 in Amend-I g ment 35 regarding the aqueous chloride solution (70 ppe chloride) at 180*F is not completely accurate. The statement (refering to results in Table IV) mentioned "No cracking has been observed in non-sensitized 304 sampler, in sensitized or non-sensitized low carbon stainless steel samples er in 308 or inconel velds." Note in Table IV thete was only a single test 9 of two sets of. duplicate U-bend specimenc in demineralized water containing 70 ppe chloride at 1BO*F. One set of specinens was in solution annealed condition and another set was in sensitized condition. The result showed one of two sensitized specimens cracked in 48 hours while the other specimens exhibited no cracking in 650 hours as in the case of the duplicate specimens in the annealed condition. It is conceivable with larger number of specimens both conditions (annealed Furthermore. ScharfsteinandBrindley(0gndsensitized)canexhibitfailure. had run similar experiments and reported both annealed and sensitized condition tracked in neutral pH solution containing 50 ppm chloride at 200*F in 48 hours. The mode of crccking uas transgranular in annealed condition and intergranular in sensitized condition. Ar. lower chloride concentration (10 ppm), the neutral solution at 165*F was slightly 9" g ei egsw = --a_e g e oe eh .e eniee - + +w we- .=e.p.-- = ,.___..m e
Mr. R. D. Young ( Novenber 20, 1968 \\ less aggressive than at 200*F. Nevertheless both conditions cracked in 312 hours. Until G.E. can produce more definitive results, this writer cannot accept the aqueous chloride solution test as a guide for site environment evaluation. Perhaps stress corrosion experiments cot. ducted by Backensto and Yuriek(7) with Type 304 stainless steel wire can be compared to the fractures in the Oyster Creek vessel. The highly stressed wires at 60,000 psi were exponed to
- 25. w/o ammonium chloride solution (pH 4.0) at voom temperature and air atmos-phere for 400 hours or until failure. No cracking was observed for annealed wir et. but sensitized wire failed intergranularly in 28 hours.
Until the cause of vessel cracking due to shop and/or site environment is clearly identified, it is advisable to examine the surfaces of external vessel components during periodic inspection or scheduled shut-downs. If no ex, amination is contemplated, then outside monitoring program for stress speci-mens of Type 308 weld overlay is advisable. Type 304 ntainless steel speci-mens in both annealed and sensitized conditiom are recommended as controls. IV. In-Reactor Intergranular Cracking Intergranular stress corrosion cranking has occurred in many water or steam-cooled nuclear reactor systems. This cracking had led to ser" ice failures of a variety of components, including thin walled (0.008 to 0.028-inch) fuel cladding,(8,9) and thick walled (0.170 to 0.50-inch) preheater pipes,(10) pres-sure vessel liners,(ll) and outlet nozzles. (II) Ccamon features of these fail-ures were. (1) the materials were austenitic stainless steels of the AISI 300 series; (2) they were exposed to acqueous or vet steam reactor coolants; and (3) they failtd intergranularly. These failures occurred in both solution-annealed steels and steels that were sensitized by carbide precipitation at grafn boundaries during velding or heat-treatment. In laboratory tests Copson et al.( } reported that a combination of oxygen contamination and crevices would produce intergranular stress corro-sion cracking of various sensitized and nonsensitized alloys (including In-conel 600 and Type 304 stainless steel) in 600*F (316*C) pH 10 water. However no cracking occurred in single U-bend specimens. Therefore, the local pH in 9 the crevice could be significantly lower than in the bulk solution. Further-more, Pickett (15) produced intergranular cracking of non-sensitized stress in 650*F low pH (3-4) water. This test solution contained 500 to 2500 soluble chromium (added as Cr0 ) ar.d 7-9 ppm oxygen. Cracking of 0.011-inch specimens 3 in these tests occurred in less than 141 hours. Similar tests in which the-soluble chrocium was added as sodium chromate to maintain a neutral pH failed to produce cracking in 2000 hour exposure. This is the reason for my recom-mondation to include crevices in gli short-r ege laboratory tests as outlired in Table III-2 of /mendment 37. \\21 ~ The results of untaxial tensile specimens in demineralized water, con-taining 100 ppm oxygen, at 550*F in fable 5 of Amce.dment 35(5) showed o e a l ~~
( Mr. R. D. Young ( -4y i Novcaber 20. 1968 intargranular cracking only occurred in the case of sensitized Type 304 and Type 304 veld overlay while annealed Type 304 and 308L veld overlay (16)exhibited no cracking. This is in agreement with recent "indings of Armijo. The latter vent on a step further. He introduced crevices at the grain boundary of a solution annealed Type 304 specimen. This was accomplished by etching the specimen in toiling HNO -Crd solution for about 12 hours to produce deep 3 ('v 0.006 inch) surface intergranular corrnsion. The specimen was then washed in boiling water and exposed to the same oxygenated watet test. Still no cracking was observed. TLe added information has convinced the writer that this type of crevice was insufficient to produce intergranular cracking in solution annealed Type 304. It is gratifying t point out that this type of crevice is similar to fine root-veld cracks. The question remains that the effects of dit. solved corrosion products and solution pH must be determined before the intergranular cracking of aus-tenitic stainless steel under stress in oxygenated and high temperature water can be understood. In the final analysis, any intergranular cracking of internal vessel components (i.e. shroud support ring) that developed under normal operating conditions is expected to be a ductile fr&cture. Brittle fracture and re-lease of small lumps of material to the coolant are not anticipated. Very truly yours, 5 , ypJ ' Metallurg,pheng Craig F.. y Division CFC:sv CC: M. V. Nevitt 1. .. Kassner . Okrent , R. A. Noland 9 e n m. = -,. - ~
1 e j ? I \\ References t 1. Letter of Craig F. Cheng, Metallurgy Division (ANL) to Laurid Porse, Div. of Reactor Licensing (AEC, Washington, D. C.), dated July 24, 1968. 2. Docket 50-219 Amendment No. 37, Oyster Creek Nuclear Power Plant, Table 111-2. April 1968. 3. Docket No. 50-219 Amendment No. 43, Oyster Creek Nuclear Power Plant, . October 1968. 4. Docket No. 50-?.19 Amet.dment No. 40, oyster Creek Fucicar Power Plant, August 1968. 5. Docket No. 50-219 Amendment No. 35, Oyster Creek rgxlear Power Plant, Marc. 1966. 6. T-R. Scharfstein and W. F. Brindley, Corrosion, p. 588t, 1958. 7. E. B. Backensto and A. N. Yurick, Cotrosion, p. 169t, 1962. 8. W. H. Arlt and S. R. Vanderberg, General Electric Co. Report GEAP-4360 (1963). 9. C. N. Spalaris, Nucleonics, 21, 41 (1963). 10. Combustion Engineering Co., Report CEND 265 (1965). 11. N. Balai, C. R. Sutton, E. A. Wimune, and R. F. Jones, Argonne National Laboratory Report ANL 7117 (1965). 12. S. R. Rideout, Savannah River Laboratory Report DP8PU 62-30-26 (1963). 13. H. R. Cupson and S. W. Dean, Effect of Contaminants on Resistance to Stress Corrosion Cracking of Ni-Cr Alloy 600 in Pressurized Water, Corrosion, 21, 1 (1965) January. 14. H. R. Copson and G. Economy, Effect of Some Environmental Conditions on Stress Corrosion Behavior of Ni-Cr-Fe Alloys in Pressurized Water, Corro-sion, 24, 55 (1968) March. \\ 15. A. E. Pickett, W. L. Pearl and M. C. Rewland, Am. Nuclear Soc. Trans. 7,
- t. 421 (1964).
16. J. S. Armijo, Corrosion, p. 319, Vol. 24, Oct. 1968. O 4 7 _v
) ( ( 09T MDBese w Prefect: Oyster Creek Unit No. 1 1@ ' EL.fq Jyle,3 : Provisional Operating License - Letter Requested f' SVt*. g. . %L,; m Rackground: The eighth Subcoensittee meeting to sensider the appliestion for on December 4, 1968.a provisional operating liconee for the Oyster Creek reac The application has been considered at three full Cosmaittee meetings. items reviewed included the following:At the Subcommittee meeting en 1. Pressure Yessel Repairs i 2. l Primary Systems Leak Detection and Procedures Whed Leaks are Detected 3 3. Technical Specifications l 4 Comparison of In-Service Inspection with Proposed M-45 Code 5. Operating Staff, Including Training, Experience Organisation, ilupport by GE. etc. 6. Radiolysis and the Advantages of Inerting 1968 Subconnaitt>se meeting:It is contemplated that the following 1. Reactor Bu11 ding Closed Cooling Water System 2. Variability of Flux Trip Point with Flow 3. Feedwater Control Valves on FWCIS 4. Subchannel Separation (Including Physical Esperation) 5. Instrumentation-General considerations 6. Auto-relief A-C Interlock and Infonmation Available to Operator f 7. Cable Tray Overloading Attached is en exce:pt from the Suarsary of the November 1968 meeting excerpt lists ther . This meeting and tha resolution of each iten. items discussed regarding the Oys DRL Reports } Since the time of the November 1968 ACES meeting, DRL has provided Repor and Supplement to Report No. 5. .5 with the pressure vessel repairs which have been made.In Report No. 5, URL In the Supplement. DEL discusses the resolution of various items and states that their r of the project is now complete. December 1968 meeting regarding: They plan to repurt to the Committee at the 1. The technical specificatione Eb ' 4. Plans to handle the probisa of radiolytic decomposition of water F 3. Sensor cables for both core spray systems being located in the 5 a. er, ~ ^ 4. Whether a Diass i fire hall has been' installed in! OfflCE ) .th t..he ttery.;tuost _.... " " " " ' " " " ' ~ " ' $URNufEb 4 4 - 4 "-- *- - - DATE> 3w AEC.318 (Rev. e-833 I' U.s sovtR=a E57 mmac orm - e-m ga
g t r i 1_tems Discuss! r { at November 1968 ACRS Meeting 6. Results of Discussio_ns Emergency Core Cooling System--resolved (more or less). a. b. Steam line isolation valves--unresolved, testing continues, a questic ~ remains on actuator operation in accident environment. k_ c. Emergency condenser isolation valves--unresolved, more next meeting. Applicant to be asked to consider a study regarding installation of a valve inside the dry well. d. Primary system leak detection--unresolved, applicant is not using all techniques availabic to him, response to a detected leak (investigate at 15 gpm) is not eatisfactory. i l Steam dryers--resolved, design satisfactory. e. j f. Inservice inspection--unresolved, more next month. Itaximum flood height--resolved, inconsequential discrepancy between g. applicant and Staff exists. 1 i h. Rod position indication--mentioned but not discussed. Radiolytic decomposition of water--unresolved, testing ac !!umbolt Bay i. continues, draft report on sources of hydrogen due about Nov. 22,196E
- j. Increased rod worths due to being left fully inserted--resolved, vendo claims condition no worse than any previously analyzed.
i l l k. Operator training and competency--unresolved, more at nexp meeting. i i 1. Reactivity anomolies--unresolved. I Technical Specifications--unresolved, more at next meeting. m. ~ f Reactor Vessel repairs--mentioned but not discussed, major item to be n. discussed at next meeting. ) Inspection of Reactor Vessel repairs--unresolved, more at next meeting. o. 1 h 6 6 E ~ ~ ^ 7.,......_+...-.,-~..-- e t
~.. - -. - g i j Corrosion surveillance of reacter vessel-- mentioned but not discussed, P. 4 q. Instrumentation and control-- mentioned but not discussed. ( 4 Cabic tray loaoings-- unresolved, more next meeting. l r. . Battery room layout-- unresolved, more next meeting. a. Fuel clad shattering-- unresolv,ed, vendor believes has traced to oxygen t. diffusion into intersticles. Auto-relief interlock-- unresolved, more next meeting. u.
- =e**
w g m _ e., ,g I I i g b I + f I a 4* 9 e .%? es g q*% 4 t I h I 4 t 9 p.s ...E l v r i 4 4 % ( 9 0 l
- "9
.e l .e '*4 mq, G . m - m--. -,,., - --,m,.- 2-- ..m r --f ,,y
oWk ,f ote 3 ms r
SUMMARY
OFSIGNIFICANTCHANGESTOTHEOkSTER' CREEK TECHNICAL SPECIFICATIONS' AS SUBMITTED IN AMENDMENT NO. 44 (1) Specification 2.1: Safety Limit - Fuel Cladding Integrity The changbs made to this specification were (1) delete reference to transients of 25 seconds and (2) express the means to demonstrate conformance with the safety limit in the event of a transient per Specification 2.1-C. ~ (2) Specification 2.3: Limiting Safety System Settings N The significant changes made to this specification con-cern items (1)a. (2)a and 6. Items (1)a and (2)a are changes in the equations for variable scram and rod block so that the LS3 levels for j each remain below the sc.fety limite given in Fig. 2.1.1. Item 6 has been changed by adding a tolerance band on the settings for the safety valves. l l (O Specification 3.1: . Protective Instrumentation I Table 3.1.1 wa9 augmented with the addition of requirements l for minimum number of Operable Instrument Channels Per Operable Trip System. { (4) Specification 3.2: _ Reactivity Control The significant change to this specification was the addi-i tion of item D which concerns Reactivity Anomalies. l (5) Specification 3.3: Reactor Coolant The significant change to this specification was the reduction in allowable primary coolant leakage as follows: unident.ified leakage: from 15 gpm to 5 gpm identified leakage from 35 gpm to 20 gpm Total leakage from 50 gpm to 25 gpm e 4 p:
.. = -. i \\ (6) Specification 3.6: Radioactive Effluents The changes in the gaseous effluent limits are due to dif ferences in the applicant's and DRL's treatment of effective stack height and interpretation of site j meteorology. The change in primary coolant resulted from the differences in accident assumptions between GE and DRI. for a steam ifne break accident particularly with regard to an iodine partition factor. (7) Specification 4.3: Reactor Coolant Substantial changes to Table 4.3.1 were made to reflect the augmented in-service inspection program proposed for Oyster Creek No. 1. Items A-2, 4, 9, 10 and 11, B-2 and 3, and all of C and D are new. In addition, Notes 2d and 3 have been added. ~ (8) Specification 4.5: Containment System This specification was rewritten to more clearly describe the containment testing program. The significant changes made to the specification are: (a) The allowable leak rate was changed from 5.0% per uay to 1.0% per day at 35 psig. (b) The test frequency was changed from a 1, 3, 5, 5, j-year program to a 1, 2, 4, 4 year frequency due to the reduction in allowable leak rate. (c) The charcoal filter efficiency requirement has j been increased from 90% to 99%. l (d) Specification L relating to the particulate filters has been added, 1 I (9) Specification 6.0: Administrative Controls The following changes-have been made to this specification. 1 /
y i s. \\ Specification Page No. Change 1.1.5 1.0-3 Added sections E, F. and G covering component failures, degradation of barriers and operating errors.
- 6. lc'2d (4) 6.1-6 Added specification regarding frequency of audits by GORE.
Fig. 6.2.2 6.1-9 Added specification requiring senior reactor operator license for shift j foreman. Added cpecification of number of people per shift. Deleted three footnotes, two of which I were redundant to Part 50; the third I had permitted acting shift foreman to possess only a reactor operator's license. 6.2A5 6.2-1 Added specification requiring written procedures for refueling operations. 6.2C 6.2-2 Added new section C requiring that instructions be.*.ssued calling for adherence to written procedures. 6.5 6.5-1 Recast to require permanent retention of records associated uith major plant changes. l /- t 5 J =}}